ML20040F653
| ML20040F653 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek, Limerick, 05000000 |
| Issue date: | 06/23/1970 |
| From: | Case E US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Morris P US ATOMIC ENERGY COMMISSION (AEC) |
| Shared Package | |
| ML20040F238 | List: |
| References | |
| FOIA-81-385 NUDOCS 8202100033 | |
| Download: ML20040F653 (10) | |
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-e 7.c 2C', uill not urite the protection end. control systc= section I[
of the t.CR*, report but vill provide n reper to D2L which cca bc
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used : a basis for developing thic section of the ACas report.
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I 4
AUG 4 1970
(
N ENCLOSURE 2 k
QUESTIONS ON PROTECTION AND CCNTROL SYSTEMS t
(For transmittal to the applicant if the areas of concern are not y
adequately covered in the SAR)
M*)
1.
In regard to the protection systems which actuate reactor trip y
H and engineered safety feature action, the following information s!
should be provided:
_: ~
N
-> f a.
A list of those systems designed and built byCenaral Electric that are identical to those of the Edwin I.11atch Nuclear Plant
[i h
(as documented in the SAR) and a discussion of any design r
differences;
[$i4 b.
A list of those systems and their suppliers that are designed
$?y and/or built by suppliers other thanGeneral Electric
~
and
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T i
c.
Identification of those features of the design which differ
_4j
_si from the criteria of IEEE 279 and the Commission's proposed
'M Tj General Design Criteria and an explanation of the reasons f
N for any differences.
1, 2.
In regard to the General Electric designed control systems,
~_
the following information should be provided:
+
n.
Identification of the major plant control systems (e.g.,
j primary te=perature control, primary water level control, 4
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_3y steam generator water level control) which are identical 31 o.
v to those in the E& sin I. Batch Nucicar Plant
- and F
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A list and a discussion of the design differences in S
those systems not identical to those used in the E&rin I. Hatch
[x:
Nuclear F1mest This discussion should include an
'g q-evaluation of the safety significance of each design change.
]
3.
State the seismic design criteria for the reactor protection
,7 system, engineered safety feature circuits, and the emergency
[
power system.
The criteria should cover (1) the capability to
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initiate a protective action during maximum peak acceleration, and (2) the capability of the engineered safety feature circuits
[
f to withstand seismic disturbances during post accident operation.
g Describe the qualification testing requirements which will be 3
i<
used to assare that the criteria are satisfied and the means by 3
3 which these requirements will be imposed on equipment suppliers.
j 4.
Describe the quality assurance procedures which apply to the f
equipment in the reactor protection system, engineered saftty
_-t feature circuits, and the emergency electric power system. This
?:5 description should include:
(a) quality assurance procedures 4,-
y used during equipment fabrication, shipment, field storage, field
/3 installation, and system co=ponent checkout; and (b) records pertaining to (a) above.
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_i, 34" AUG 4 1970
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e 5.
Submit the criteria and their bases which establish the minimum
[V requirements for preserving the independence of redundant reactor
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protection systems, engineered safety festare systems and Class IE*
[f:
f Electrical Systems through physical arrangement and separation and ie rs assure minimum availability during any
- design basis event. The I
A. -
submittal should include a discussion of the ad=inistrative res-j pensibility and control to be provided to assure compliance with h!
these criteria during the design and installation of these systems.
h The criteria and bases for the installation of electrical cable
?
for these systems should, as a minimu=, address:
- }f-2
_j a.
Cable derating.
l b.
Cable routing in containment, penetration areas, cable Y.
_t spreading rooms, control rooms and other congested or
_{
hostile areas.
I
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c.
Sharing of cable trays with non-safety related cables or I
witn cables of the same system or other systems.
((
d.
Fire detection and protection in the areas where these
-h y,
cables are installed.
}^}
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Cable and cable tray markings.
f.
Spacing of wiring and components in control boards, panels, h
h and relay racks.
x
- Class IE electrical systems and design basis events are defined in the at Proposed IEEE Griteria for Class IE Electrical Systems for Nuclear Power kI Generating Stations (IEEE-308).
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AUG 4E
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6.** State the design criteria for reactor protection system and
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engineered safety feature related electrical and mechanical 5cb equipment located in the primary containment or elsewhere in N
R:
the plant which take into account the potential effects of
[j:<
i radiation on these components due to either normal operation a
or accident conditions (superimposed on long-term normal R
N,$
operation). Describe the analysis and testing performed to g
eru verify compliance with these design criteria.
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as 4{?.
7.** Identify all safety related equipment and co=ponents (e.g.,
+
motors, cables, filters, pump seals) located in the primary i[.,
h containment which are required to be operable during and subse-W-
$E g'.<'
quent to a loss of coolant or a steamline break accident. Des-cribe the qualifications tests which have been or will be performed d4 y
on each of these items to insure their availability in a combined 47D high temperature, pressure, and humidity environment.
g
&4 8.
State the criteria which have been established to assure that J!
~p loss of the air conditioning and/or ventilation system will 3g i,,j not adversely affect operability of safety related control and g
electrical equipment located in the control room and other x
2 M
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- Ihese questions relate to the engineered safety feature chapter of the
- p.
SAR and should be forwarded to the applicant with other questions con-Y cerning that chapter.
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W lU N N M d 9_ M D W I " T @ N ~
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-['i 4tJG 4 1970 cdi,,
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~Ib equipment rooms.
Describe the analysis performed to identify ys 1%
the worst case environment (e.g., temperature, humidity).
State
-#^J 2d' the limiting condition with regard to temperature that would g%
require reactor shutdown, and how this was determined.
Describe
~
any testing (factory and/or onsite) which has been or will be
['
5
~4 performed to confirm satisfactory operability of control and N{>:
electrical equipment under extreme environmental conditions.
-f7 un I
Ib 9.
Describe how reactor protection system and engineered safety z.n
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Ec '
equipment will be physically identified as safety related if-equipment in the plant.
aAW 10.
Describe the method for periodic testing of engineered safety
_A,
m
-4 feature actuation to show it to be consistent with IEEE 279.
E.
We interpret IEEE 279 to require for engineered safety feature
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actuation the same high degree of on-line testability required
)
B for the reactor trip system.
C Rb '
Provide a description of the instrumentation systens included
?
11.
-d
- .gl in your design for remote monitoring of post-accident conditicac s.
within the primary containment. Provide an analysis to show y;
l that these systems are adequate over the full spectrum of M
M
=1 1
postulated accidents.
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MICLosURE.3-
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- 6 5
QCTESTIONS ON ENERCE ICT 7012ER FOR LDERICE CENERATDtG ST E
y e
1, ;.
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s Provide a drawing that shows the aerial trant tssion line rooting a
1.
both to and from the 220 kV and the 500 kV substations.
. 1 w
y
~.
Discuss the analyses to be performed to show that maither loss f[u I
2.
of a 1mnit of this station nor tha' loss of the largest 3==Har
}
b, unit oc the grid vill negate the ability to provide offsite hi 9
pomr to this station.
g, S*(
Clarify the apparent inconsistency between Figurn 8.4.1 and i
3.
Provide a description of the substation Figure 8.3.1 of the TSML.
fonders cutmeted to each of the 13.2 kY secanAarias of the start-g up transforwars as *nM ented on Figure 8.4.1 of the PSAL Describe the 4-c power system for both the 220 kV and the 500 kV 4.
substa teen.
r Provide the design critaria and tafomarinn mains the onsite 5.
electrical pwer aptems as follows-t b
~
The percentage of the continuous ratings of each diesel h,
A.
generator that the engineared safety features electrical l
The continoons rating is defined as loads will require.
that continuous load which will permit suppliar guaranteed 4
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operation at a 95% availability with an '=rmaat maine===
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,ariod.
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B.
he 2000-hour and the 30 stincte diesel generator overload 49
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a tinga.
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C.
A description of the rg. ired prototype taats and pre-jj
~
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opera *4nnal tests to assic= the adequacy of design and O
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reliability of these largE capacity diesel generators
+%
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which are new to nuclear service.
Mst g
D.
Criteria for usinnining independence and separatico of Q'
wp the ecoling water system, fuel oil system, and the control bN system for each of the diesel generator smita.
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OO.
E.
A table to illustrate the automatic and nur2nal sequencing M
.ht w
of *ian angineered safety feature electrical loeds on
?;Lj
(
M, -
each of the diesel generators. Include the time of each h
l
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event and the electrical load, fror. the initiating signal
?j
.s.
x to a fully loaded c<= die 4 =.
P'
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F.
A description and analysis to de=xxistrate that cooling water h
M to the diesel generators and other energency loads during g
Di loss of offsite power is allowable. Your proposed design tg with twosnergency service water pu=ps, one being supplied k
0 T4 $
m mz
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+
5 ys
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25 o
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y Y-AUG i EXi Enclosura.3 3
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from a nuit 1 amargency bus and the other from a Unit 2 16 35
. boa, aoea not a,,sar eo -as-3 the, ingle s.ttura orttert g
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b 6.
Provide the following information concerning the 24,125, ad -
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32 l 250 voit d-c power systmas 4
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A.
Se bases for =4M of each of the d-e battarias. -
g B.
A description of tha venti 1=rian system **i14H for the f.g
.,e battery rooms. Include bow independence and separation
- jj sig are maineaina.
+,
Y' C.
With your arrangement of two 125 volt d-c batteries located F
v:
"D in a - room it can be assumed that a =<ni m event 5
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all loads being supplied by both'betteries. to be
-y lost. Provide an analysis to justify that the safety of
.R^
N_
s the plant would not be jeopardised by this e,ent.
4 ri s
7.
Page 8.6-3 of the PSAR statas W%-t necessary for safe t*
xr shutdown of the plant is supplied with control power from redan-3+
a dant 125 volt battery sources. Zoes of one source would not
-FF prevent safe shutdown of the plant." Provide a more detailed
- h,
- k description to assure that a true split-bus design wi1I. be main n4w in achieving the above objective.
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